This invention relates generally to nuclear reactors and more particularly to a design analysis method that permits operation of a boiling water nuclear reactor in an expanded region of the power/core flow map.
A typical boiling water reactor (BWR) includes a pressure vessel containing a nuclear fuel core immersed in circulating coolant, i.e., water, which removes heat from the nuclear fuel. The water is boiled to generate steam for driving a steam turbine-generator for generating electric power. The steam is then condensed and the water is returned to the pressure vessel in a closed loop system. Piping circuits carry steam to the turbines and carry recirculated water or feedwater back to the pressure vessel that contains the nuclear fuel.
The BWR includes several conventional closed-loop control systems that control various individual operations of the BWR in response to demands. For example a control rod drive control system (CRDCS) controls the position of the control rods within the reactor core and thereby controls the rod density within the core which determines the reactivity therein, and which in turn determines the output power of the reactor core. A recirculation flow control system (RFCS) controls core flow rate, which changes the steam/water relationship in the core and can be used to change the output power of the reactor core. These two control systems work in conjunction with each other to control, at any given point in time, the output power of the reactor core. A turbine control system (TCS) controls steam flow from the BWR to the turbine based on pressure regulation or load demand.
The operation of these systems, as well as other BWR control systems, is controlled utilizing various monitoring parameters of the BWR. Some monitoring parameters include core flow and flow rate effected by the RFCS, reactor system pressure, which is the pressure of the steam discharged from the pressure vessel to the turbine that can be measured at the reactor dome or at the inlet to the turbine, neutron flux or core power, feedwater temperature and flow rate, steam flow rate provided to the turbine and various status indications of the BWR systems. Many monitoring parameters are measured directly, while others, such as core thermal power, are calculated using measured parameters. Outputs from the sensors and calculated parameters are input to an emergency protection system to assure safe shutdown of the plant, isolating the reactor from the outside environment, if necessary, and preventing the reactor core from overheating during any emergency event.
To meet regulatory licensing guidelines, the thermal output of the reactor is limited as the percentage of maximum core flow decreases. A line characterized by this percent of thermal power output versus percent of core flow defines the upper boundary of the reactor safe operating domain. Some reactors have been licensed to operate with increased thermal power output (up-rated) with an upper boundary line characterized by the point of 100 percent original rated power and 75 percent of rated core flow. This upper boundary line constrains operation at the uprated power to a significantly smaller range of core flow and reduces flexibility during startup and at full power.
In one aspect, a computerized method for expanding the operating domain of a boiling water nuclear reactor that permits safe operation of the reactor at core flows lower than normal operating parameters is provided. The operating domain is characterized by a map of the reactor thermal power and core flow. The computerized method includes determining by computer simulation a load line characteristic that is elevated over normal operating parameters and that increases reactor performance over normal operating parameters, performing safety evaluations by computer simulation at the elevated load line to determine compliance with safety design parameters, and performing operational evaluations by computer simulation at the elevated load line.
In another aspect, a system for controlling a boiling water nuclear reactor is provided. The system includes a computer configured to determine a set of operating characteristics for the reactor in an upper operating region above 100 percent of a rated core thermal power by simulation, evaluate an expected performance of the reactor throughout the upper operating region by simulation, and determine limits for the reactor that are to be observed within the upper operating region by simulation.